Systems and methods for fast molten salt reactor fuel-salt preparation

ABSTRACT

The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

CROSS-REFERENCE TO RELATED APPLICATIONS

This application is a continuation of U.S. patent application Ser. No.17/527,862, filed Nov. 16, 2021, which is a continuation-in-part (CIP)of International Application No. PCT/US2020/032902, filed May 14, 2020,which claims priority to, and the benefit of, U.S. patent applicationSer. No. 16/415,668, filed May 17, 2019, and U.S. patent applicationSer. No. 16/415,692, filed May 17, 2019, each of which is incorporatedherein in its entirety by reference.

TECHNOLOGICAL FIELD

The present disclosure relates generally to methods and systems forprocessing pelletized-form light water reactor spent nuclear fuel intofluoride-based or chloride-based molten salt reactor fuel.

BACKGROUND

Nuclear fuel that has been irradiated in a nuclear reactor is generallyreferred to as spent nuclear fuel. Such spent nuclear fuel is generallynot reused or recycled but instead is most often stored onsite inspecially-design pools in the vicinity of the nuclear reactor in whichsuch nuclear fuel was used. Aged spent nuclear fuel, having undergonesignificant decay so that it doesn't produce significant heat, may alsobe stored in dry casks on pads at the reactor site in which it wasproduced, at decommissioned reactor sites, and/or at other approvedsites pending disposal at a permanent disposal facility.

According to the U.S. Energy Information Administration, as of 2013,there were more than 70,000 metric tons of spent nuclear fuel stored atsites within the United States(https://www.eia.gov/nuclear/spent_fuel/). Such spent nuclear fuel willbe lethal to humans for thousands of years, requiring its storage meetstringent requirements and close monitoring. Further, while spentnuclear fuel storage has proven to be reasonably safe to date, the riskremains that a large fire, explosion, terrorist attack, plane crash, oraccident could occur that damages a spent fuel pool and/or dry caskstorage of such spent fuel.

Therefore, it would be desirable to have a system, apparatus and/ormethod that takes into account at least some of the issues discussedabove, as well as possibly other issues, and yields an improvedsolution.

BRIEF SUMMARY

Exemplary implementations of the present disclosure are generallydirected to methods and processes involving light water power reactorspent nuclear pellet-form fuel, extracted from fuel assemblies, or, rodscontaining such spent nuclear pellet-form fuel, separating fuel rodcladding from the pellets, retaining substantially all spent fuelpellets and fragments thereof, and processing to halide salt powdersuitable for use, perhaps with enrichment, to fuel for a molten saltreactor.

Exemplary implementations of the present disclosure are generallydirected to methods and systems resulting in a substantially completeautomated process for making new fuel for a molten salt reactor andwhich, consequently, reduce the inventory of spent fuel at commercialpower reactor sites. In certain of such implementations, virtually noneof the contents of the spent fuel are removed, virtually all of thespent fuel and contents are converted to molten salt fuel, andessentially all reprocessed/reprocessed fuel will be capable ofgenerating power, substantially without the generation of additionalwaste, resulting in what could potentially be near-continuous reductionof currently-stored spent fuel inventory over time, with the consequentvolumetric reduction of highly radioactive waste to stable or low-levelforms.

In exemplary implementations of the present disclosure, processes arediscussed for converting uranium oxides and plutonium oxides (as used)to uranium and plutonium chlorides.

Implementations of the present disclosure are generally directed tomethods and systems for preparation of a fluoride-based salt fuel for athermal molten salt reactor (TMSR).

Further implementations of the present disclosure are generally directedto methods and systems for preparation of a chloride-based salt fuel fora fast molten salt reactor (FMSR).

Implementations of the present disclosure are generally directed tomethods and systems for retaining essentially all fuel materials in aspent fuel recycling process to substantially fully close the nuclearfuel cycle, thereby virtually eliminating a nuclear waste stream ofrejected material. Highly radioactive fission waste and trans-uraniumelements, e.g., Actinides in the spent fuel are retained in the processand new fuel product.

Implementations of the present disclosure are generally directed tomethods and systems for allowing fission waste products and actinides inspent nuclear fuel to be burned to stable forms and incorporate saltfuel for enrichment in order to overcome the negative power effects ofthe fission product waste.

Certain implementations of the present disclosure are generally directedto methods and systems for reducing spent nuclear fuel and constituentsto fluoride fuel salt, including calculated enrichment (U235 or Pu239).Uranium or plutonium is added to the spent fuel before or duringfluoridation. The final product is a dry fuel salt powder.

Other implementations of the present disclosure are generally directedto methods and systems for producing chloride fuel salt by moltenchloride reduction of spent nuclear fuel and constituents, includingcalculated enrichment (U235 or Pu239). Uranium or plutonium is addedduring chlorination. The final product is a dry fuel salt powder. In avariation of this implementation, molten salt fuel product may be pouredinto canisters, stored, and later inductively heated to liquid and usedin the molten salt reactor.

Implementations of the present disclosure are generally directed tomethods and systems that avoid chemical separation of spent fuelconstituents into separate streams (which would generally result in morestorage of highly radioactive waste) or additional chemical processes torejoin spent fuel groups.

According to one example implementation, a facility and process designof the present disclosure allows operators to process spent nuclear fuelstored at nuclear power plants. The process, once put into practice,essentially closes the nuclear fuel cycle, by reducing the amount ofnuclear spent fuel stored at plant sites, and molten salt reactor fuelproduced in accordance with this disclosure is intended to be used inadvanced molten salt reactors to ultimately generate electricity, hotwater, etc. According to another exemplary implementation, a processfacility is provided that is a hardened, secure, limited-accessfacility, configured to accept and contain highly radioactive materialspent nuclear fuel, and specifically designed to house apparatusmachinery and attendant support systems for continuous spent fuelprocesses, including spent fuel container ingress and processing spentfuel to a final packaged product, and egress of such final product. Thisincludes shielding against radiation, remote robotic operations and safehandling, and the exclusion of personnel from processing area.

In another example implementation, a facility for conducting a processof the present disclosure is located on site within a secured perimeter,together with a spent nuclear fuel storage facility and a molten saltreactor.

According to a further exemplary implementation of the presentdisclosure, processes are disclosed capable of utilizing virtually allwater-reactor ceramic spent nuclear fuel consisting of uranium oxide,lanthanide series elements, fission product metals and non-metals,actinide series elements, i.e., substantially all of the material foundin spent nuclear fuel, for the production of halide salt fuel. Theseprocesses avoid “wet” chemical separation, since element (nuclide)constituents are not chemically separated. Production of fluoride saltis by a dry process, whereas, the process for the production of chloridesalt is a liquid emersion conversion to salt. There are effectively noleft-over waste products or waste streams.

According to one further exemplary implementation of the presentdisclosure, fuel assemblies containing an array of fuel tubes arealigned horizontally on a rod puller disassembly table, and spent fuelpellets are removed from tubes, or, “pins,” by laser slitting of thefuel tubes, opening the tubes, and mechanically removing any spent fuelpellets and fragments that remain adhered to tubes. Cleaned fuel tubes,channels, and assembly end pieces (non-fuel), are set aside forrecycling.

According to one further exemplary implementation of the presentdisclosure, spent fuel pellets and fuel pieces are processed for aspecific molten salt reactor type, i.e., fluorinated salt for a“thermal” reactor and chlorinated salt for a “fast” reactor.Specifically, fuel for a thermal molten salt reactor is processed to afluoride salt by ultimate reduction and fluorination of uranium and itsassociated fuel constituents, and fuel for a fast molten salt reactor isprocessed to a chloride salt by ultimate reduction and chlorination ofuranium and its associated fuel constituents.

According to one further exemplary implementation of the presentdisclosure, spent fuel pellet removal from fuel pins is performed in aclosed atmosphere to prevent release of dust and to capture gases backinto the process. In such implementations, fuel for a molten saltreactor proceeds to one of two parallel path, specifically: (a) a pathwherein at least one rotating calciner is used in a fuel fluorinationpreparation process, which receives extracted spent fuel pelletsdirectly (“Option A”), or (b) a path wherein chlorinated fuelpreparation spent fuel pellets are ball-milled in an enclosed atmosphereto collect dust and gases, which are recycled back into the process, andspent fuel pellets are milled to coarse granular feed to the molten salturanium/plutonium oxide reduction tank (“Option B”).

According to a further exemplary implementation of the presentdisclosure, spent fuel pellets undergoing Option A, i.e., thefluorination process, may be enriched by adding U235 powder to the spentfuel before fluorination, to predetermined specifications, to providesufficient fissionable material in the final product. This enrichmentcould be low enriched uranium including high assay-low enriched uranium(e.g., HA-LEU, <20% enrichment). In one implementation, spent fuel andconstituents are reduced to fluoride salt powder in a rotary calciner,and the powder may be enriched as necessary with fluorinated uranium-235or plutonium-239 in order to support molten salt reactor operation.

According to yet a further exemplary implementation of the presentdisclosure, spent fuel undergoing Option B, a chlorination process, maybe enriched with U235 or Pu239 powder added to spent fuel salt at auranium/plutonium oxide reduction tank, to the desired specifications,to thereby provide enough fissionable material in the final product.This enrichment could be low enriched uranium including high assay-lowenriched uranium (e.g., HA-LEU, <20% enrichment) or plutonium, or mixedoxide (MOX) fuel. In one implementation, spent fuel and constituents arereduced to chloride salt fuel by immersion in a molten chloride saltbath. Molten chloride fuel salt may then be enriched, in the eventinitial and subsequent enrichments will be required, with uranium-235,plutonium-239, or MOX (uranium and plutonium), added to the oxidereduction tanks, in order to support molten salt reactor operation.

Proper sizing of tanks and equipment, including, without limitation,multiple oxide reduction tanks being in parallel, and multiple mixingand adjustment tanks being in parallel, allows for enrichment andchemical analysis, and required tank separation and reactivitymonitoring. Proper sizing ensures ample volume for mole-fractionconcentration specifications. Mixing tanks provide for sampling,adjustment, and content certification. Separation, and partitioning ofboth sets of tanks, a first group (oxide reduction tanks) and a secondgroup (mixing and adjustment tanks), ensures sub-criticality during theentire process. Molten chloride salt fuel is solidified and milled topowder in the process. Additional chemical processes, such as fluidizedbeds or small chemical reactors, may be employed to process fuel dust orvolatile constituents to stable form and to subsequently rejoin spentfuel process streams. Final-product molten salt fuels are provided inpowder form, sample-tested, and certified.

According to a further exemplary implementation of the presentdisclosure, in the case of chlorinated molten salt, initial calculationsdetermine how much granulated spent fuel will be added to each oxidereduction tank of molten alkali or alkali earth chloride by mass andconcentration of free chloride. The number and size of molten chloridereaction tanks necessary for continuous process operation to producechlorinated salt-fuel will be determined, in part, by the rate of spentfuel pellet production and milling, spent fuel chlorination to salt, andsafety considerations.

According to a further exemplary implementation of the presentdisclosure, spent fuel gases evolved during the Options A and Cfluorination and Option B chlorination processes, are collected by afluidized bed, converted to fluorinated and chlorinated fuel salts,respectively, and returned to their separate fuel salts.

According to yet another exemplary implementation of the presentdisclosure, chlorinated fuel salt is pumped from at least one reactingtank, by its own pump when it and its discharge isolation valve areselected. Fuel salt is pumped to a common header, with a selector valvefor each to admit fuel salt to at least one cooling tray. Piping andvalves from reacting tanks to cooling trays are maintained above saltmelt temperature by redundant and remotely replaceable heating elementjackets, to prevent salt fuel solidification.

According to a still further exemplary implementation of the presentdisclosure, chlorinated fuel salt in the molten state is introduced intocooling trays, cooled by chilled water, designed with multiple parallel,but separate rows, each surrounded by cooling coils to remove heat fromthe molten salt and cause it to solidify. The cooling trays are in astacked array, with adequate space between trays for addition of moltensalt and removal of solidified. Actual configuration and groupings of anumber of trays into one array, depends on the: (a) movement ofganged-arrays; (b) tray loading from the pump-out of the mixing andadjustment tanks; (c) stacked array movement to cooling stations; (d)stacked array movement to the ball mill feed table (not shown); (e)solidified salt fuel removal and deposit on the ball mill feed table;and (f) salt fuel milling.

According to a yet further exemplary implementation of the presentdisclosure, solidified and cooled chlorinated fuel salt bars, or“sticks,” are removed from their molds, after solidification. Thesesticks are collected and fed to a ball mill and fine mill, which isenclosed to retain process dust, for processing to powder, and milled tospecifications. The number of cooling trays will be sufficient tosupport continuous feed to the milling operation and chlorinated fuelsalt powder system demand.

According to a yet another exemplary implementation of the presentdisclosure, fuel salt product from Option A, is collected from thecalciner and milled to powder product specifications, and fuel saltproduct from Option B is collected from fuel salt mold, ball-milled, andfurther milled to a fine powder. An alternate implementation of Option Bmethod provides certified fuel salt directly to individual storagecanisters, which are inductively heated to liquid for use in a moltensalt reactor.

According to a yet another exemplary implementation of the presentdisclosure, end products of the fluorination process (Option A), and endproducts of the chlorination process (Option B) include their respectivesalt fuels and are milled to specifications and then collected incontainers on carts, sealed, and transported by cart for direct use orstorage. Container geometry and amount of salt fuel product are sized toprevent criticality in stored arrays. Fission product nuclides and otherneutron absorbing barriers help assure adequate margin to criticality inall potential concentrations.

Another exemplary implementation of the present disclosure in theproduction of fuel for a thermal molten salt reactor by: providing spentfuel pellets; processing the spent fuel pellets and fuel pieces into afluoride fuel salt by ultimate oxidation, reduction, and fluorination ofuranium and its associated fuel constituents in a generally continuousprogression, wherein the processing produces water vapor; and filtering,condensing, and removing the water vapor produced during the reductionand fluorination operations.

Another exemplary implementation of the present disclosure provides amethod for producing fuel for a thermal molten salt reactor including:(a) providing fuel assemblies containing an array of fuel tubes alignedhorizontally on a rod puller disassembly table and removing fuel pelletsfrom the tubes; (b) processing the spent fuel pellets and fuel piecesinto a fluoride salt by ultimate oxidation, reduction, and fluorinationof uranium and its associated fuel constituents; and (c) filtering(including condensing and removing) the water vapor formed during thereduction and fluorination operations. In an additional exemplaryimplementation, the method could further include enriching the granularspent fuel salt with U235, and if further desired, fluorinating theU235-enriched granular spent fuel salt or plutonium in a calciner rotarykiln or fluidized bed.

In some exemplary implementations, a method for producing fuel for afast molten salt reactor is provided which includes providing fuelassemblies removing fuel pellets containing uranium from the fuelassemblies and granulating the fuel pellets into granular spent fuelsalt for processing feed in a semi-voided atmosphere using a ball mill,roller mill, or chopping mill, and processing the granular spent fuelsalt into chloride salt by ultimate reduction and chlorination of theuranium and associated fuel constituents of the uranium. The methodfurther includes enriching the granular spent fuel salt with U235,Pu239, or MOX, chlorinating the enriched granular spent fuel salt toyield molten chloride salt fuel using anhydrous HCl and halide saltreduction, and then analyzing, adjusting, and certifying the moltenchloride salt fuel for end use in a molten salt reactor. Additionally,the method includes pumping the molten chloride salt fuel to stackedarrays of cooling trays or canisters and cooling the molten chloridesalt fuel to yield solid salt fuel bars, sticks, or canister solid formsand milling the solidified molten chloride salt fuel to predeterminedspecifications for the fast molten salt reactor.

A method for producing fuel for a fast molten salt reactor, the methodincluding providing fuel assemblies and removing fuel pellets containinguranium and all spent fuel constituents, from the fuel assemblies andgranulating the fuel pellets in a semi-voided atmosphere using a ballmill, roller mill, or chopping mill, for process feed to thechlorination process. The granular spent fuel salt is processed intochloride salt by ultimate reduction and chlorination of the uranium andassociated fuel constituents chloride salt solution. Reduction may occurusing a strong reducing agent, preferably a chloride-containing reducingagent, such as anhydrous hydrogen chloride (AHCl). The granular spentfuel salt is enriched with U235, Pu239, or MOX, and the enrichedgranular spent fuel salt is chlorinated to yield molten chloride saltfuel using AHCl halide salt reduction. The molten chloride salt fuel isanalyzed, adjusted, and certified for end use in a molten salt reactor.This implementation also includes pumping the molten chloride salt fuelto stacked arrays of cooling trays or canisters and cooling the moltenchloride salt fuel to yield solid salt fuel bars, sticks, or canistersolid form, and milling the solidified molten chloride salt fuel topredetermined specifications for the fast molten salt reactor.

Non-limiting example approximate temperatures, times, gasconcentrations, materials used to construct the apparatus, and otherparameters which are expected to be used are shown in the drawings.

More specifically, implementations of the present contemplate afull-size facility enclosing the methods and processes for processingcommercial light water reactor spent nuclear fuel to final product fuelsalt product for a molten salt power reactor. Equipment and machinery inthe facility receive spent fuel assemblies and deliver them to a rodpulling table, and disassemble them from their support elements, namely,into separated, individual cladding fuel rods containing raw spent fuel.The fuel rods are slit and/or sliced axially along substantially theirentire length by laser. Care is taken to prevent the laser from cuttingthrough the spent fuel, burning, or fusing pellets. In oneimplementation, the laser simultaneously cuts opposite sides of each rodinto semi-cylindrical halves, thereby exposing fuel pellets when the twohalves of the fuel rod are separated. Prior to the fuel rod claddingsections being removed from the process, mechanical brushes sweep theinside surfaces of such sections lengthwise in order to recover all ofthe spent fuel pellets and pieces thereof, given such fuel pellets andpieces may exhibit various forms, from being generally intact, i.e.,cylindrical, to broken and deformed shapes, indicated by previous cyclesof operating history and subsequent handling. Gaseous constituents ofthe spent fuel are collected during disassembly and the conversionprocesses discussed herein and are processed within a fluidized bed intohalides for recovery.

In one exemplary implementation, thermal reactor salt fuel requires lowneutron energy (thermal energy) for thermal fission to occur, wherebyneutrons immediately begin to lose energy quickly after they areproduced from fission to continue the process of thermal fission. Thisis achieved by conversion of spent fuel to salt fuel of light masselemental salt, whereby light mass elemental metals of beryllium orlithium, for example, and fluorine, form salt fuel effecting thermalfission.

In another exemplary implementation, during fluorination, light masselemental metal hydrides of beryllium or lithium, for example, form saltfuel effecting thermal fission by reduction of spent fuel to salt fueland oxidation of hydride light mass metals to salts, whereby duringoperation, thermal molten salt reactor neutrons quickly lose energy tothermal energy, after they are produced from fission, to continue theprocess of thermal fission.

In one exemplary implementation, fast reactor salt fuel requires highneutron energy for fast fission to occur, and such energy is desired tobe greater than the threshold for fast neutron energy, whereby neutronsretain enough energy after they are produced from fission to continuethe process of fast fission. This is achieved by conversion of spentfuel to salt fuel of heavier mass elemental salt, whereby heavy masselemental metals of potassium, zirconium, or zinc, for example, andhalides of chlorine, bromine or iodine, form salt fuel effecting fastfission.

In another implementation, during halogenation, heavy mass elementalmetal hydrides of zirconium, molybdenum, or tin, for example, form saltfuel effecting fast fission by reduction of spent fuel to salt fuel andoxidation of hydride heavy mass metals to salts, whereby duringoperation, fast molten salt reactor neutrons retain energy well abovefast neutron threshold energy after they are produced from fission tocontinue the process of fast fission.

According to a further exemplary implementation of the presentdisclosure, processes are disclosed capable of utilizing virtually allwater-reactor ceramic spent nuclear fuel consisting of uranium oxide,lanthanide rare-earth series elements, fission products, actinide serieselements, i.e., substantially all of the material found in spent nuclearfuel, for the production of halide salt fuel. These processes avoid“aqueous wet” chemical separation, since fission product waste andactinide constituents are not chemically separated from nuclear fuelmaterial. Production of fluoride salt fuel and chloride salt fuel, andwaste fission products and actinides, is a liquid immersion conversionin molten halide salt fuel. There are effectively no left-over wasteproducts or waste streams.

According to another exemplary implementation of the present disclosure,fuel assemblies containing an array of fuel tubes are alignedhorizontally in a disassembly table box for breakdown of the fuelassembly into individual fuel rods, and relocation of fuel rods to apellet extraction grid table. Spent fuel pellets are removed from tubes,by horizontally laser cutting the fuel tube on both sides the entirelength of the tube, opening the tubes, and mechanically removing anyspent fuel pellets and fragments that remain adhered to tubes. Cleanedfuel tubes, channels, and assembly grid spacers and end pieces, are setaside for recycling.

In a further exemplary implementation of the present disclosure, spentfuel pellet removal from fuel pins or tubes, is performed in a closedatmosphere to prevent release of dust and gases, which are recovered,converted to salt fuel and returned back into the process. In suchimplementations, fuel for a molten salt reactor proceeds to one of twoparallel paths, specifically: (a) spent fuel pellets are milled in aclosed atmosphere, to a powder in preparation for conversion to thermalmolten salt reactor fluoride salt fuel as feed to the uranium/plutoniumoxide reduction tank, or (b) spent fuel pellets are milled in a closedatmosphere, to a powder in preparation for conversion to fast moltensalt reactor chloride salt fuel as feed to the uranium/plutonium oxidereduction tank.

According to a further exemplary implementation of the presentdisclosure, spent fuel and constituents are reduced to fluoride saltfuel by immersion in a molten fluoride salt bath.

In another exemplary implementation, a method further includes enrichingthe powdered spent fuel salt with U235, Pu239, or MOX fuel, chlorinatingthe enriched powdered spent salt fuel to yield molten chloride salt fuelusing anhydrous hydrogen chloride (AHCl) reduction, in molten chloridesalt, and then analyzing, adjusting, and certifying the molten chloridesalt fuel for end use in a molten salt reactor. The method furtherincludes ultimate reduction of uranium oxide and constituent oxidesusing metal hydrides. Generally, in exemplary implementations of thepresent disclosure processes for conversion of powdered spent nuclearfuel, used fuel, to molten salt reactor salt fuel begins with a startingbase bath of molten halide salt, or a mixture of halide salts as themolten medium to dissolve all spent fuel constituents. Particular acidsof the halides e.g., hydrogen—fluoride, chloride, bromide, or iodide,may be used for halogenation of uranium, plutonium, fission products andactinides by “fluorination,” “chlorination,” “bromination,” or“iodination” of powdered spent nuclear fuel, converting it to “saltfuel.” Generally, halide salt, e.g., sodium chloride or potassiumchloride, and anhydrous hydrogen chloride are used for spent fuelconversion to chloride salt fuel. This is necessary to initialize andmaintain a continuity of salt fuel physical and nuclear characteristics.

Other exemplary implementations of the present disclosure include amethod of processing spent nuclear fuel pellets into molten salt reactorfuel, the method comprising milling the spent nuclear fuel pellets intospent nuclear fuel powder and feeding to a halide forming process,wherein the halide includes at least one of chloride, bromide, andiodide, and processing the spent nuclear fuel powder into halide salt byultimate reduction; halide forming of the uranium and associated fuelconstituents in a halide salt solution comprised of a bath of selectedmetal hydride salts; enriching the halide salt; and halogenating theenriched halide salt to yield molten halide salt fuel.

Exemplary implementations of the methods may include that the step ofprocessing the spent nuclear fuel powder into halide salt occurs byreacting the halide salt with at least one of anhydrous hydrogen halideand metal hydride, which could occur in an oxide reduction tank, and incertain exemplary implementations may include the anhydrous hydrogenhalide and/or metal hydride being provided via a sparger in an oxidereduction tank. Enrichment of the halide salt may also take place in theoxide reduction tank, and hydrogen may be created, converted to water,with the water being generally continuously removed from the oxidereaction tank.

Certain exemplary implementations of the present disclosure includeplacing the molten halide fuel salt in a canister, covering the moltenhalide fuel salt with argon gas, and sealing the canister with themolten halide fuel salt and argon gas therein.

Other exemplary implementations of the present disclosure include amethod of processing spent nuclear fuel pellets into molten salt reactorfuel, the method comprising milling the spent nuclear fuel pellets intospent nuclear fuel powder and feeding to a fluoride forming process;processing the spent nuclear fuel powder into fluoride salt by ultimatereduction; fluoride forming of the uranium and associated spent nuclearfuel powder constituents in a fluoride salt solution comprised of a bathof selected metal hydride salts; enriching the fluoride salt; andfluorinating the enriched fluoride salt to yield molten fluoride saltfuel.

Certain exemplary implementations of the present disclosure may includethe step of processing the spent nuclear fuel powder into fluoride saltincludes reacting the fluoride salt with anhydrous hydrogen fluoride,which could be in an oxide reduction tank and potentially through use ofa sparger in the oxide reduction tank. Also, the fluoride salt could beenriched in the oxide reduction tank.

Exemplary implementations of the present disclosure may include placingthe molten fluoride fuel salt in a canister; covering the moltenfluoride fuel salt with argon gas; and sealing the canister with themolten fluoride fuel salt and argon gas therein.

In some exemplary implementations, spent fuel gasses from the spentnuclear fuel powder are collected by a fluidized bed of chemical reactorand converted to fluorinated fuel salts.

Other exemplary implementations of the present disclosure include asystem for processing spent nuclear fuel pellets into molten saltreactor fuel, the system including a mill configured for milling thespent nuclear fuel pellets into spent nuclear fuel powder and an oxidereduction tank configured for receipt of the spent nuclear fuel powderand for containing a process for forming the spent nuclear fuel powderinto halide salt by ultimate reduction; halide forming of the uraniumand associated spent nuclear fuel powder constituents in a halide saltsolution comprised of a bath of selected metal hydride salts; enrichmentof the halide salt; and halogenating the enriched halide salt to yieldmolten halide salt fuel.

In still other exemplary implementations of the present disclosure, asystem is provided for processing spent nuclear fuel pellets into moltensalt reactor fuel and includes a mill configured for milling the spentnuclear fuel pellets into spent nuclear fuel powder and an oxidereduction tank configured for receipt of the spent nuclear fuel powderand for containing: a process for forming the spent nuclear fuel powderinto fluoride salt by ultimate reduction; fluoride forming of theuranium and associated spent nuclear fuel powder constituents in afluoride salt solution comprised of a bath of selected metal hydridesalts; enrichment of the fluoride salt; and fluorination of the enrichedfluoride salt to yield molten fluoride salt fuel.

The features, functions and advantages discussed herein may be achievedindependently in various exemplary implementations or may be combined inyet other exemplary implementations further details of which may be seenwith reference to the following description and drawings.

BRIEF DESCRIPTION OF THE DRAWINGS

Having thus described exemplary implementations of the disclosure ingeneral terms, reference will now be made to the accompanying drawings,which are not necessarily drawn to scale, and wherein: FIG. 1schematically illustrates methods and systems according to exemplaryimplementations of the present disclosure for use in processing spentnuclear fuel into molten salt reactor fuel, and more specifically, afluoride-based salt fuel for a thermal molten salt reactor (TMSR);

FIG. 2 schematically illustrates methods and systems according to anexemplary implementation of the present disclosure for use of a moltensalt preparation in processing spent nuclear fuel into chloride orfluoride fuel salt, and more specifically, the receiving of pellet-formspent fuel, which has been milled to pulverized powdered form anduranium/plutonium oxide reduction tank, wherein the purpose is to removeoxygen and prevent production of other oxides, results in the removal ofoxygen as water and conversion of generated hydrogen to water, obviatingthe need for using costly catalysts or chemicals not readily available,and upon oxygen and hydrogen being removed from the fuel salt,substantially the only byproduct is water;

FIG. 3 schematically illustrates the mixing and adjustment tank,illustrates methods and systems according to exemplary implementationsof the present disclosure for use of a molten salt sampling, adjustment,and certification, in processing spent nuclear fuel into chloride orfluoride fuel salt;

FIG. 4 schematically illustrates in plan view an exemplary methods andimplementations of the present disclosure for use of a molten saltpreparation in processing spent nuclear fuel into chloride or fluoridefuel salt, and more specifically, typical separation for operation andcritical-safe parallel arrangements of molten salt spent fuel oxidereduction and mixing tanks, including oxide reduction tanks and mixingand adjustment tanks and their relative orientation and physicalseparation, and physical separation by neutron absorbing panels,generally boron composites, to ensure subcriticality between parallelarrangements of oxide reduction tanks, and the same neutron absorbingpanels between parallel arrangements of mixing and adjustment tanks toensure subcriticality, and including an oxide reduction tank dischargeheader discharges to the mixing and adjustment tank without usingnozzles;

FIG. 5 schematically illustrates methods and systems according toexemplary implementations of the present disclosure for use of a moltensalt preparation in processing spent nuclear fuel into chloride orfluoride fuel salt, and more specifically, salt mold cooling trays;

FIG. 6 schematically illustrates methods and systems according toexemplary implementations of the present disclosure for use of a moltensalt preparation in processing spent nuclear fuel into chloride orfluoride fuel salt, and more specifically, a salt mold cooling tray,including a top cover, cooling molds, and heating and cooling coils;

FIG. 7 schematically illustrates a method and system according to anexemplary implementation of the present disclosure, including, a processfor thermal molten salt reactor (TMSR) fuel-salt preparation;

FIG. 8 schematically illustrates a method and system according to anexemplary implementation of the present disclosure, including, processesfor fast molten salt reactor (FMSR) and thermal molten salt reactor(TMSR) fuel-salt preparation;

FIG. 9 schematically illustrates methods and systems according to anexemplary implementation of the present disclosure for use of a moltensalt preparation in processing spent nuclear fuel into chloride orfluoride fuel salt; and

FIG. 10 schematically illustrates a site on which components of a systemaccording to an example implementation of the present disclosure may belocated.

DETAILED DESCRIPTION

Some implementations of the present disclosure will now be describedmore fully hereinafter with reference to the accompanying drawings, inwhich some, but not all variations of the disclosure are shown. Indeed,variations of the disclosure may be embodied in many different forms andshould not be construed as limited to the examples set forth herein;rather, these are provided so that this disclosure will be thorough andcomplete and will fully convey the scope of the disclosure to thoseskilled in the art.

As used herein, “and/or” means any one or more of the items in the listjoined by “and/or.” As an example, “x and/or” means an element of thethree-element set, e.g., [(x), (y), (x, y)]. Additionally, as usedherein, the terms “exemplary” and “example” mean in context as servingas a non-limiting example, instance, illustration, or circumstance.Moreover, as used herein, the term “for example,” or, “e.g.,” introducesa list of one or more non-limiting examples, instances, illustrations,or circumstances.

Exemplary implementations in accordance with the present disclosure aredescribed with reference to systems and/or methods, such as in thecontext of processing spent nuclear fuel. Further, for example,reference is made herein to values of or relationships betweencomponents, parameters, properties, variables or the like. These andother similar values or relationships are absolute or approximate toaccount for variations that may occur, such as those due to engineeringtolerances or the like. Like reference numerals refer to like elementsthroughout.

Of note, the disclosures set forth in Conversion of Oxide to Metal orChloride, by Sakamura, et al., Organization Central Research Instituteof the Electric Power Industry (CRIEPI), Japan, and Effect of MeltComposition on the Reaction of Uranium Dioxide with Hydrogen Chloride inMolten Alkali Chlorides, by Volkovich, et al., Ural State TechnicalUniversity, Russia and the entirety of both of the foregoing documentsare incorporated herein by reference.

Further incorporated by reference in their entirety are the followingdocuments: Processing of Used Nuclear Fuel, World Nuclear Association,(updated June 2018),https://world-nuclear.org/informaiton-library/nuclear-fuel-cycle/fuel-recycling/processing-of-used-nuclear-fuel.aspx);Recycling Nuclear Fuel: The French Do It, Why Can't Oui?, Dec. 28, 2007,The Heritage Foundation(https://www.heritage.org/environment/commentary/recycling-nuclear-fuel-the-french-do-it-why-cant-oui;Recycling Process of Defective Aged Uranium Dioxide Pellets, FatahMernache, et al, published online Aug. 12, 2015, Journal of NuclearScience and Technology, Vol 53, Issue 6; Engineering Design of aVoloxidizer with a Double Reactor for the Hull Separation of SpentNuclear Fuel Rods, Young-Hwan Kim, et al Korea Atomic Energy ResearchInstitute, Science and Technology of Nuclear Installations, Vol 2017,Article ID 985; Oxidation of UO2 Fuel Pellets in Air At 503 and 543 KStudied Using X Ray Photoelectron Spectroscopy and X Ray Diffraction, P.A. Tempest et al, Journal of Nuclear Materials February 1988; The HighBurnup Structure in Nuclear Fuel, Vincenzo V. Rondinella et al, EuropeanCommission, Joint Research Centre, Institute for Transuranium ElementsGermany, Materials Today, December 2010, Vol 13, No 12; UraniumTetrafluoride, IBILABS International Bio-Analytical Industries, Inc.Aug. 7, 2016; Uranium Tetrafluoride, Wikipedia Ref Journal of theAmerican Chemical Society, 1969; Hydrofluoric Acid Corrosion Study ofHigh-Alloy Materials, P. E. Osborne et al, ORNL, UT Battelle, LLC forDOE, August 2002; and “Inconel 600”, Spec sheet FSA, Shanghai FengquSuperalloy Co, Ltd. Mar. 13, 2019.

Additionally, incorporated by reference in their entirety are thefollowing patent documents: GB 803258; GB 1171257; GB 2536857; JP11231091; KR 20060035917A; KR 20090089091A; KR 2009010 109237A; KR20090109238A; KR 20110034347A; US 2013/0266112A1; WO 2017/158335A1; US2011/0286570A1; U.S. Pat. Nos. 9,767,926; 4,062,923; and 6,251,310.Further incorporated by reference in its entirety are the documents:“Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuelin CANDU Reactors—I: DUPIC Fuel Fabrication Cost, by Hangbox Choi, WonLi Ko, and Myung Seung Yang, Korea Atomic Energy Research Institute,Nuclear Technology, Vol. 134, May 2001; Proceedings of the 16thInternational Conference on Nuclear Engineering ICONE 16, 2006/2008“Second Generation Experimental Equipment Design to Support VeloxidationTesting At INL”; World Journal of Nuclear Science and Technology, 2015“Reduction Kinetics of Uranium Trioxide to Uranium Dioxide UsingHydrogen.”

Briefly, FIGS. 1 and 7 illustrate an exemplary implementation of asystem including one or more processes for thermal molten salt reactor(TMSR) fuel-salt preparation, and FIG. 8 illustrates another exemplaryimplementation of the present disclosure, namely, a system including oneor more processes for fast molten salt reactor (FMSR) fuel-saltpreparation.

Methods and Systems for Calciner Fluoride Fuel Salt Preparation(“OptionA”)

FIGS. 1 and 7 illustrate an exemplary implementation of a system,generally 100, according to one exemplary implementation of the presentdisclosure for calciner fluoride fuel salt preparation in the productionof thermal molten salt reactor fuel salt. An externally heated andcooled calciner apparatus provides for a continuous process forconverting spent fuel UO₂ pellets (which have been previously removedfrom the fuel cladding) at elevated temperatures to UF₄ crystals/powderusing a rotating cylinder through which the pellets advance, and using acounter-flow of oxygen (via the center of an axial support tube of anintegrated helical auger) for oxidizing, then reducing in a concurrentflow of hydrogen gas, then fluorinating in a concurrent flow of HF gas(the gasses being introduced via the central pipes contained in the axisof the calciner.) The calciner apparatus has sealing mechanisms at bothends to prevent any of the gases or particulates from entering thesurrounding facility's atmosphere. The pellets are loaded through thesesealing mechanisms at one end of the calciner apparatus, and the UF₄continuously exits through such sealing mechanisms at the other end ofthe calciner apparatus.

In some aspects, for example, the system 100 process begins with spentfuel pellets being recovered from fuel rod cladding (not shown) and fedinto a rotating calciner, generally, 106. In an exemplaryimplementation, two calciners, one for each of two lines, could be used.As shown in FIG. 1 , an axial cross-section of a calciner 106 depictsthe construction thereof and the gas flow there-through, shown by arrows112. The direction of process flow through the calciner, is from left toright as indicated by arrows, in the axial-cross section. Sealed entriesand exits to and from the calciner prevent gases escaping outside thesystem, while allowing entry of spent fuel pellets and exit of fluoridesalt. (See U.S. Pat. No. 7,824,640, to Pitts, incorporated herein in itsentirety by reference). All gases used and recycled in the calciner willbe filtered to exclude unwanted elements and particles from exiting withthe product. (See U.S. Pat. No. 4,666,684, to Pitts, incorporated hereinin its entirety by reference). Proceeding through the calciner 106,there are three process subdivisions designated by three radialcross-sections, or zones, generally 106A, 106B, and 106C, (FIG. 1 )showing the particular gas flow of each section and desired product.

Excess gases leave the calciner 106 by negative pressure to externalfilters (not shown). The externally-heated calciner rotates slowly,heating pellets to approximately 500° C. for a period of time, which inone non-limiting example could be approximately 1 to 3 hours. Section106A includes a fixed integral helical auger, the direction of rotationbeing indicated, as viewed in the direction of gas flow from right toleft in FIG. 1 .

Calciner 106 dimensions, in one non-limiting example, could beapproximately 15 to 30 inches in outside diameter, and axial section Acould be approximately 10 to 20 feet in length. Axial sections 106B and106C, in one non-limiting example, could be approximately 5 to 10 feetin length each. Sensors, which in some non-limiting examples may beembedded or attached wireless micro-sensors, generally 114, are shown inthe calciner casing 116 and serve to monitor process parameters such astemperatures, pressures, material and added constituents flow rates,radiation, gases, and/or other measurable process details.

One center conduit, or pipe, 118 extends the entire length of thecalciner 106, which has a plug in the mid length of the pipe to preventthe mixing of the oxygen and hydrogen gasses. Oxygen (an oxidizingagent) enters at the left (as shown in FIG. 1 ), as pellets enter thecalciner 106, and the oxygen exits from the pipe 118 into the calcinerinterior 120, starting the oxidation of pellets. Curved arrows 112 onthe axial section 106A indicate oxygen flow from the pipe 118 to thecalciner interior 120. Spent fuel pellets, generally 124, are indicatedin section 106A at the helical auger 126 as are also mixing vanes 128.During this part of the process in calciner 106, UO₂ (uranium dioxide)spent fuel pellets are oxidized to various oxides of uranium, whichcauses the pellets to disintegrate because of expansion duringoxidation. All other constituents of spent fuel are contained andoxidized in this section.

A smaller center pipe 130 enters from the right end of calciner 106 anddoes not penetrate the full length of the calciner, but insteadterminates at the start of axial section B. A baffle 131 will be used toreduce the mixing of the oxidizing gas and the reducing gas, at theappropriate spot axially, in the calciner, but will still allowadvancement of the product through the calciner. (See U.S. Pat. No.3,969,477, incorporated herein in its entirety by reference). Onenon-limiting exemplary location of baffle 131 is shown in FIG. 1 .Baffle 131 is between the oxygen flow and the hydrogen gas flow in thecalciner 106. The center pipe 130 supplies hydrogen gas for the secondpart of the process that takes place in the calciner 106, as shown bycurved arrows 112 indicating outflow into the calciner main volume. Thehydrogen gas is a reducing agent and flows from the center pipe 130 intothe calciner interior 120, in section 106B. During this part of theprocess, various oxides of uranium are reduced to UO₂, and theconsistency of the spent fuel pellets has been changed from a generallypellet form to coarse powder, shown in section 106B at the helical auger126 and mixing vanes 128. Virtually the only effluent is water vapor,which is condensed during filtration of the recirculating gasses. Allother constituents of spent fuel are contained and reduced during thispart of the process.

The final process converts UO₂ to UF₄ (uranium tetrafluoride). Hydrogengas continues to flow through the smaller center pipe exiting into thecalciner section 106B, as described previously; then, HF (hydrogenfluoride) gas enters into the larger annular pipe 118 at the right endof the calciner 106 as shown in FIG. 1 , and exits into the maininterior, or, body, 120 of the calciner 106 at the beginning of axialsection 106C, with arrows 132 indicating direction of flow. During thispart of the process, uranium and virtually all other constituents in thespent fuel, fission products, rare earths, and actinides are allsubstantially fluoridated. The resultant coarse powder product is shownin cross-section 106C at the helical auger and mixing vanes 128.

Design and construction of the calciner apparatus 106 may include anysuitable manufacturing techniques, including without limitation,application of 3D printing in order to use heat and corrosion resistantmaterials to create a durable internal design of calciner 106.

Calciner 106 includes instruments and sensors for the measurement ofpressure, temperature, gas concentration, gas flow, and material flow,which can be accomplished by many, perhaps hundreds, of wirelessimbedded micro sensors 114, which are monitored in real-time by computersystems and artificial intelligence applications to maintain safety ofoperation and to provide continuous improvement of the process. Thesensors 114 may be built into the calciner apparatus 106 during the 3Dprinting process.

The calciner apparatus 106 keeps radioactive particles contained toprevent contamination of the surrounding facility, and calcinerapparatus 106 generally produces only relatively small volumes ofcondensed liquid waste water, which will require specialized disposal.Operation of calciner apparatus 106 is more easily automated foroperation on a 24/7 basis and is potentially less-expensive to operateover its lifetime than other types of processing. The design of theprocess using calciner apparatus 106 is scalable for increased capacity,as well as lending itself to be standardized for replication, so thatmultiple units can be used for backup purposes and/or to increasefacility capacity.

The conversion gases used in calciner apparatus 106 are carried by inertgases such as helium or argon, which are recycled. Water vapor generatedduring processing is condensed and removed from the process on acontinuous basis. Gases exit the calciner apparatus 106 at each of thesealed ends to the filtering and replenishment equipment. Therecirculated gases are filtered to remove elements not desired in theend product.

During the first stage of this process, as shown in section 106A of FIG.1 , the pellets 124 are exposed to a counter-flow of oxidizing gas,shown by arrow 125, such as oxygen, to covert the UO₂ to variouscombinations of higher oxides of uranium, which increases the volume ofthe pellets up to 30%, potentially causing them to fracture. As shown inthe section 106A, the motion of the rotating cylinder 138, integralhelical auger 128 blades, with small shelves, or ledges, 140 to lift thepellets 124, provides friction between the pellets, and small impactforces experienced by the pellets hasten the oxidation process, whichitself expands and fragments the pellets 124 further. This ultimatelyresults in powdered oxides of uranium. The diameter of the calciner 106,in one exemplary implementation, could be in the range of approximately15-30 inches.

The second stage 106B of the calciner 106 process is to the right of aclosure 127 in conduit 118 (FIG. 1 ) and exposes the oxide powders ofuranium to a flow of reducing gas such as hydrogen, converting thevarious uranium oxide powders to UO₂. FIG. 1 shows the designs ofsection 106B, helical auger 128, and the axial gas supply channel 130.

The third stage 106C of the calciner 106 process shown in section 106Cexposes the UO₂ powder to fluoridizing gaseous HF, which produces UF₄ ina crystalline/powder form for use in lithium fluoride molten salt-basedreactors, for example. Section 106C shows the auger 128 and central pathof the HF gas. Such H₂ and HF gases then exit conduit 119 to filtering,which includes condensing and removing of the water vapor formed duringthe reduction and fluorination operations.

The UF₄ exits the process in a manner which prevents leaking of gases tothe atmosphere. A mechanism will be provided to seal the end of thecalciner, so that the gases generated will be contained, and the productwill exit cooled and ready for the next operation. The product issampled, tested, and certified for shipment. The UF₄ is automaticallyplaced in containers, which are automatically sealed and cooled, andthen stored for delivery to the customer. (In order to provide moreuniform particle sizes than can perhaps be produced in the calciner 106,as the product exits the calciner 106, a subsequent milling operationfor milling to powder to desired specifications may be used.) The thirdstage of the process shown in section 106C exposes the UO₂ powder tofluoridizing gaseous HF, which produces UF₄ in a crystalline/powder formfor use in lithium fluoride salt-based reactors, for example. Section106C shows the auger 128 and central path of the HF gas.

In an exemplary implementation of the present disclosure, a method isillustrated in FIG. 7 for producing fuel for a thermal molten saltreactor, the implementation of the method including:

a. providing fuel assemblies containing an array of fuel tubes arealigned horizontally on the rod puller disassembly table 270 (FIG. 9 ),and fuel pellets are removed from tubes;

b. processing the spent fuel pellets and fuel pieces into a fluoridesalt by ultimate oxidation, reduction and fluorination of uranium andits associated fuel constituents; and

c. filtering (including condensing and removing) the water vapor formedduring the reduction and fluorination operations.

Another exemplary implementation of such method could include, ifdesired and as shown in FIG. 7 , fluorinating the U235-enriched granularspent fuel salt or plutonium in a calciner rotary kiln or fluidized bed,and if additionally desired, enriching the granular spent fuel salt withU235.

In an exemplary implementation, because both the reduction of the oxidesof uranium to uranium dioxide and the conversion of uranium dioxide touranium tetrafluoride are exothermic, the calciner includes bothexternal heating and cooling apparatus (not shown) over most of itslength.

In an exemplary implementation, the temperature of conversion of uraniumdioxide to uranium tetrafluoride in HF gas is to be maintained above 400deg C. during and after the conversion is completed, to prevent theundesired formation of volatile uranium hexafluoride, which will occurif it is cooled below 400 deg C. in the presence of HF gas. Therefore,the uranium tetrafluoride must exit through the sealed end of thecalciner above 400 deg C. and then cooled to ambient temperature. Thisrequires a counter flow of argon in the exit sealing mechanism of thecalciner as cooling proceeds. Toward this end, the sealing and transfermechanism for the fuel product is to be configured with sufficientcooling capacity. The calciner is configured to reduce the likelihood ofthe oxygen and hydrogen used in processing from being too close togetherin the oxidation and reduction steps in the calciner. Although at leastone baffle 131 is used, it may be desirable to use multiple baffles,with the introduction of positive pressure inert gasses such as argon,between them, to prevent the mixing of oxygen and hydrogen during theprocess. Such inert gas can be introduced into the calciner through apipe (not shown) placed axially in the auger 126, extending from theentrance end of the calciner to the baffle area.

In exemplary implementation, Option A may include, if desired, the spentnuclear fuel being generally permanently stored, then processed intospent fuel salt, and the spent fuel salt used in a thermal molten saltreactor, all on a single site having a secured perimeter.

Non-limiting example approximate temperatures, times, gasconcentrations, materials used to construct the apparatus, and otherparameters which are expected to be used are shown in the drawings.

Methods and Systems for Chloride Fuel Salt Preparation (“Option B”)

FIGS. 2-7 and 9 illustrate an exemplary implementation of a system,generally 200, or portions thereof, according to one example of thepresent disclosure for chloride fuel salt preparation in the productionof fast molten salt reactor fuel salt.

The process 200 begins after the spent fuel pellets 124 recovered fromcladding in a manner as discussed above, being fed into a ball mill 202(FIG. 9 ), and pulverized to a granular form. Gases are recovered fromthe initial disassembly, from the ball mill 202, and from one or moreenclosed conveyors (not shown), routing granulated spent fuel to the(FIG. 2 ) oxide reduction tanks 210. The oxide reduction tanks are thefirst tanks in line of the process 200 to treat granular/pulverizedspent nuclear fuel. Spent fuel is reduced using a strong reducing agent,preferably a chloride-containing reducing agent, such as anhydroushydrogen chloride (AHCl) addition through a tank sparger 212 at thebottom of the tank 210. A small excess of chloride with molten chloridefuel salt ensures enough free chloride to produce chloride salt fuel.The process produces water vapor and hydrogen which are continuouslyremoved by blower extraction and condensation, and glow plugs (notshown) ensure hydrogen and oxygen gases are burned to water product.This process completes the goal of removing oxygen from all oxides inthe salt fuel. Automated and dip sampling configuration, and densityprobes, while provided, are not shown. Gases are collected into afluidized bed (not shown) for chlorination and recycling back into themain process. Raw granulated spent fuel is routed from the ball mills202 by the enclosed conveyor to parallel oxide reduction tanks 210containing molten salt. Granulated spent fuel is conveyed in a closedsystem, to the oxide reduction tank hopper 216.

A tank 220 containing molten chloride salt maintained, in onenon-limiting example, at approximately (30-50) degrees C. (80-120degrees F.) above the melting point of the halide salt (molten alkali oralkali earth chloride) melting point estimated to be 500 C (930 F). Themelting point of the molten salt is variable with the amount andconsistency of alkali and alkali-earth chlorides, and with the amount ofspent fuel added to the mix. Nominal density of spent fuel salt chlorideis expected to be 3.0 g/cc, depending on concentration (mol %). It isanticipated salt fuel for the fast molten salt reactor will requiresignificant enrichment. This enrichment will be performed with additionof U235, Pu239, or MOX fuel. At an estimated beginning 30 mol % uraniumchloride and plutonium-chloride, the balance being fission productchlorides and actinide chlorides (5-10) mol %, the remaining mixcontains free molten salt at (60-65) mole %.

FIG. 3 shows the fuel salt mixing and adjustment tank 220, second inline of an exemplary implementation of the process, receives fuel saltin a hopper 223 from the oxide reduction tank 210. Both tanks 210, 220(FIGS. 2 and 3 ) have automated sampling, and pump recirculationdistribution headers (not shown) internal to the tanks. Tanks 210, 220(FIGS. 2 and 3 ) are sized and configured to prevent a criticality(critical-safe) in the tank as pulverized spent fuel is added andenriched with U235, Pu239, or MOX fuel, to high assay low enricheduranium (HA-LEU) at less than approximately 20% enrichment. Both tanks210, 220 have the capability to receive salt, spent fuel, orenrichments; however, tank 220 will normally receive only saltreplenishment as needed. The enrichment is necessitated in fueling andoperation of a fast molten salt reactor. Tanks 210 and 220, in onenon-limiting example, have approximate estimated dimensions of 10 feetin height by 16 feet front to back and 10 inches wide and is capable ofprocessing approximately 1000 gallons. Tanks 220, in one exemplaryimplementation, are constructed integrally with an outside tank (notshown) having leak detection between the inside and outside tanks.Outside tank dimensions allow for insulation, multiple electric heateraccess points, recessed instrument enclosures, and accesses to each.

The tanks 210, 220 are instrumented with dip sample points (not shown)for automatic and/or manual sampling and analysis. This capabilityconfirms independent on-line sampling that a receiving-mixing tank'scontents are fully chlorinated to the extent possible (uranium, fissionproducts, lanthanides, and actinides), i.e., substantially the entireinventory of spent nuclear fuel nuclides. A density probe 221 and manualliquid density measurement generated therefrom confirm whether the spentfuel salt density is at the expected density nominally (3.0-4.0) g/cm³(kg/m³), molten alkali or alkali earth chloride density, no othercontent, is approximately (1.6 g/cm³). The contents of the oxidereduction tank 210 (FIG. 2 ), and mixing and adjustment tank 220 will beprocessed further when sample analyses are confirmed. It is estimatedreaction processing time is in an exemplary implementation 8 hours,including enrichment and sample confirmation, for the oxide reductiontank 210, and 4 hours for the mixing and adjustment tank 220. Full rangegamma and neutron nuclear instruments, generally 224, provide continuousmonitoring, trending, and alarming (counts/second) and rate of change.In one implementation, oxide reduction tank 210 size and configurationrequire four equally spaced instruments over the height and depth ofeach tank. A blower and chiller 226 combination removes water from tank210. An anhydrous hydrogen chloride cylinder and compressor, generally228, supply sparger arrangement 212. Salt mixers (e.g., screw-typemixers) 222 set at different depths, and front to back of the tank,ensure sufficient mixing of each tank. Additionally, FIG. 2 is anexemplary implementation wherein tank screw pump 218 is shown. Tankscrew pump 218 is connected to tank 210 via conduit 210A, and an inletvalve 218A is provided proximate the inlet of pump, and an outlet valve218B is provided proximate the outlet of pump 218. A conduit 218Cconnects valve 218B to a discharge valve 219 and to a tank pumprecirculation isolation valve 229 connected to molten fuel salt mixingtank 220.

In an exemplary implementation shown in FIG. 3 , a tank screw pump 218is connected to tank 220 via conduit 220A, and an inlet valve 218A isprovided proximate the inlet of pump, and an outlet valve 218B isprovided proximate the outlet of pump 218. A conduit 218C connects valve218B to a tank header valve 217 and to a tank pump recirculationisolation valve 225.

In FIG. 4 , a representation of tanks 220 containing molten salt spentfuel are shown in a plan view in a side-by-side relationship, and, as inthe case of all the drawings herein, are not shown to scale. This viewrepresents the general size and configuration for both the oxidereduction tanks 210 (side by side) and mixing and adjustment tanks 220(side by side). More specifically, FIG. 4 shows an implementationwherein six tanks 220 are shown. Accompanying apparatus and equipmentand configurations used in connection with the tanks 220, are not shown.

In the basic process flow (FIG. 9 ), it could be considered that thefirst oxide reduction tank 210 would pump out to the first mixing andadjustment tank, the second oxide tank, the second mixing tank, andcontinue this sequence until all oxide reduction tanks have pumped outto their respective mixing and adjustment tanks. The tanks, FIG. 4 ,also include the tank pump discharge header and nozzles, which are onlylocated on the mixing and adjustment tank 220 pump out header. Spacedbetween tanks are encased boron slabs, or, dividers, or encasements,230. Boron encasements 230 positioned between all fuel salt tanks 220prevent nuclear criticality communication between the array of oxidereduction tanks in close proximity, and between the array of mixing andadjustment tanks in close proximity. Each group is considered herein asone subcritical assembly group, generally 266. Boron dividers 230 arebackup defense in depth against possible criticality.

In an exemplary implementation, equipment is selected for durability andreliability. Two channels of electric “jacketed heaters” (not shown) arefitted to piping, valves and pumps ensure salt fuel in piping andequipment is of a high enough temperature to remain liquid and willflow. The heater channels are monitored, alarmed, and component failureidentified if such a failure occurs. If sections of piping are allowedto cool where molten salt is solidified, heaters can be activated tore-melt the fuel salt. Instrumentation and automated functions are fullyalarmed and continuously communicated to a control center. Diagnosticprotocols help operators identify system interruptions or pointsrequiring repair. All components on tanks and transfer piping arepreferably accessible and capable of remote repair after steps are takento isolate failed components from the system. Multiple independentreceiving-mixing tanks and transfer equipment ensures a continuoussupply of fuel salt in operation, including in the event of a systemfailure.

In an exemplary implementation, fuel salt preparation is begun withintroduction of chloride salts of alkali metals or alkaline earth metals(e.g., NaCl, KCl, MgCl₂, CaCl₂)), typically in crystalline form, andusually a mixture of two or more salts to a tank. Heaters (e.g.,electrical heating elements) 231 are energized to melt the salt tomolten state and maintain temperature well above melting point.Pulverized-granulated spent nuclear fuel is taken from the ball mill 202and carried by enclosed conveyor to the tank hopper 223 and depositedvia hopper isolation valve 227 into the oxide reduction tank, and openisolation valve 227 (FIG. 2 ). Spent fuel addition, regulated bylimiting size and speed of the enclosed conveyor, ensures tanktemperature is maintained within predetermined specifications andsufficient mixing and reaction of tank contents occurs. All mixers andthe tank pump are turned on to initiate mixing, recirculation bypumping, and consistency of fuel salt in the tank and pump dischargelines. In an exemplary implementation, tank size and screw pump capacityare regulated to allow 4-6 hours for mixing and sampling, densityrecording and analysis before a tank is ready to be discharged.

FIGS. 5 and 6 show an exemplary implementation in plan and elevationalviews of a molten fuel salt collection tray, generally 240, consistingof the tray cover 254, fuel salt collection molds 256 atop heating andcooling elements 242, in a generally checkerboard pattern of heatingelements 234A and cooling elements 234B. Insulation between elementsalong the sides and bottom of the tray prevents solidification duringpouring and minimizes cooling time after solidification. Morespecifically, FIG. 5 is a plan view of salt mold cooling tray 240 withthe tray cover and cooling molds removed.

In an exemplary implementation, salt mold cooling trays 240 (FIG. 9 )are positioned and held in a stacked array of 8-10 trays, with spacingbetween the trays being sufficient to allow for removal of the coolingmolds and cover as one assembly. Stacked arrays are tracked together bya revolving drive (not shown) which moves one stacked and cooled arraygroup, to the ball mill feed table 250. At the table 250, the one-piecemolds of each tray, in a particular group, are removed and upended oroverturned to deposit solid “bars” of fuel salt. Each tray's mold isremoved and returned to its position before the next mold is removed.After an entire array group has been emptied, it is returned in turn tobe refilled with molten salt. Tray molds can be a non-stick surface,with salt fuel contraction during cooling, thereby facilitating solidsalt fuel removal. The metal molds may be connected side-to-side andlaterally supported to ensure tray strength and versatility. Solid salt“bars” are gathered to the side of the turning table and are generallyorganized lengthwise on a moving conveyor and fed into 2-3 coarse ballmills 202A (FIG. 9 ). Product fuel salt from the ball mill is furtherconveyed to a fine mill, such as a Fitz mill 252 for sizing, sampling,certification, and packaging for protection against environmentalconditions. More specifically, FIG. 6 illustrates a salt mold coolingtray front view, including top cover 254, cooling molds 256, heating andcooling elements 234A, 234B, which, in one implementation, could becoils.

In an alternate implementation, molten salt fuel may be stored as acontiguous solid in canisters and subcritical arrays. This processinvolves preparation of chloride fuel salt in the aforementionedreceiving and mixing tanks 220, sampling and certification of tanks, andtransfer by screw pump to a “critical safe” steel canister (not shown),set aside for cooling. Canisters are transported and stored, in“critical safe” arrays. Facilities using “solid salt” canisters areequipped to remotely handle and inductively heat each canister to formliquid fuel salt for addition to their molten salt reactors.

In FIG. 4 , molten salt spent fuel receiving and mixing tanks 220 areshown in a side-by-side representative top view, but not to scale. Morespecifically, FIG. 4 shows an implementation wherein sixreceiving-mixing tanks 220 are side-by-side, including a tank pumpdischarge header, generally 260, and encased boron encasements 230(sized for inner tank side dimension area). Boron encasements 230provide structural stability and protection from damage and arepositioned between fuel salt tanks 220 to prevent nuclear criticalitycommunication between tanks in a given array of tanks in closeproximity, such array being considered herein as one subcriticalassembly group, generally 266. Boron dividers 230 are backup defense, indepth, against possible criticality.

As shown in an exemplary implementation in FIG. 2 , the oxide reductiontank is the first tank in the process, wherein a system and process areillustrated which reduces uranium and plutonium oxides to chlorides.After this process, the contents of the reduction tank 210 are pumpedover to the mixing and adjustment tank. In one implementation, anisolation valve is provided on the loading hopper. In an exemplaryimplementation wherein only chloride salts are desired, the salts areprepared in the oxide reduction tank, and, then pulverized, granulatedspent fuel is added to the oxide reduction tank, the temperatureincrease is noted, and oxide reduction is begun by the sparging actionof anhydrous hydrogen chloride. After an allowance of time necessary formixing and water and hydrogen gas removal, toward complete reduction,the contents are pumped to the mixing and adjustment tank for finalanalysis, certification, and then pumped over to cooling trays. Theteaching of the present disclosure includes reducing all oxides,removing oxygen entirely and preventing production of other oxides,ensures an authentic chloride salt fuel, when oxygen is removed from thefuel salt. The result is that substantially the only byproduct from thisimplementation of the present disclosure is water, which is collectedfor sampling and released.

In an exemplary implementation of the present disclosure, a method isillustrated in FIG. 8 for producing fuel for a fast molten salt reactor,the implementation of the method including:

a. providing fuel assemblies, removing fuel pellets containing uraniumand all spent fuel constituents, from the fuel assemblies;

b. granulating the fuel pellets in a semi-voided atmosphere using a ballmill, roller mill, or chopping mill, for process feed to thechlorination process;

c. processing the granular spent fuel salt into chloride salt byultimate reduction and chlorination of the uranium and associated fuelconstituents chloride salt solution, by anhydrous hydrogen chloride(AHCl);

d. enriching the granular spent fuel salt with U235, Pu239, or MOX;

e. chlorinating the enriched granular spent fuel salt to yield moltenchloride salt fuel using AHCl halide salt reduction;

f. analyzing, adjusting, and certifying the molten chloride salt fuelfor end use in a molten salt reactor;

g. pumping the molten chloride salt fuel to stacked arrays of coolingtrays or canisters and cooling the molten chloride salt fuel to yieldsolid salt fuel bars, sticks, or canister solid form; and

h. milling the solidified molten chloride salt fuel to predeterminedspecifications for the fast molten salt reactor.

In exemplary implementations, Option B may include, if desired, thespent nuclear fuel being generally permanently stored, then processedinto spent fuel salt, and the spent fuel salt used in a fast molten saltreactor, all on a single site having a secured perimeter.

Non-limiting example approximate temperatures, times, gasconcentrations, materials used to construct the apparatus, and otherparameters which are expected to be used are shown in the drawings.

FIG. 10 schematically illustrates a site on which components of a system1000 according to an example implementation of the present disclosuremay be located. The system 1000 may include a site 1002 within a securedperimeter 1004, and a limited-access facility 1006 on the site 1002. Asdescribed above, the system 1000 may include a number of components1008, and at least some of these components 1008 of the system 1000 maybe located within the limited-access facility 1006. The system 1000 mayalso include a spent nuclear fuel storage facility 1010 located on thesite, and a molten salt reactor 1012 located on the site 1002.

In other exemplary implementations of producing chloride salt fuel usingsystem 200, additional reduction may be achieved by addition of metalhydrides. Generally, in exemplary implementations of the presentdisclosure the processes follow that described above for Option Bprocesses for conversion of powdered spent nuclear fuel, used fuel, tomolten salt reactor salt fuel begins with a starting base bath of moltenhalide salt, or a mixture of halide salts as the molten medium todissolve all spent fuel constituents. Particular acids of the halidese.g., hydrogen—fluoride, chloride, bromide, or iodide, may be used forhalogenation of uranium, plutonium, fission products and actinides by“fluorination,” “chlorination,” “bromination,” or “iodination” ofpowdered spent nuclear fuel, converting it to “salt fuel.” Generally,halide salt e.g., sodium chloride or potassium chloride, and anhydroushydrogen chloride are used for spent fuel conversion to chloride saltfuel. This is necessary to initialize and maintain a continuity of saltfuel physical and nuclear characteristics.

As discussed above, the oxide reduction tanks are the first tanks inline of the process to treat pulverized/powdered spent nuclear fuel.Spent fuel is reduced using a strong reducing agent, preferably achloride containing reducing agent, such as anhydrous hydrogen chloride(AHCl) addition through a tank sparger 212 at the bottom of the tank210. Additional reduction may be achieved by addition of metal hydrides.A small excess of chloride with molten chloride fuel salt ensures enoughfree chloride to produce chloride salt fuel. The reduction of uraniumoxide, plutonium and substantially all spent nuclear fuel constituentsproduces hydrogen and oxygen forming water vapor and are continuouslyremoved by blower extraction and condensation. Generally continuousremoval of water during oxide reduction is essential to maintain anacidic balance continuously, during spent fuel reduction. And, inanother exemplary implementation, numerous glow plugs (not shown) ensurehydrogen gas and oxygen are burned to water product, are placed near thetop of the tank interior, have redundant power supplies and glow plugfailure monitoring. All ancillary equipment for production of watervapor from hydrogen and oxygen, and removal of water from the tanks, isdesigned with significant margin in excess of the maximum expectedprocess generation rate. This process completes the goal of removingoxygen from all oxides or hydroxides, in the salt fuel. Automated anddip sampling configuration, and density probes are also provided (notshown). Gases are collected into a fluidized bed or small chemicalreactor (not shown) for chlorination and recycling back into the mainprocess. Raw powdered spent fuel is routed from the ball mills 202 bythe enclosed conveyor to parallel oxide reduction tanks 210 containingmolten salt. Powdered spent fuel is conveyed in a closed system, to theoxide reduction tank hopper 216. Tank 220 containing molten chloridesalt is maintained, in one non-limiting example, at approximately(30-50) degrees C. (80-120 degrees F.) above the melting point of thehalide salt (molten alkali chloride) melting point estimated to be 600 C(1048 F). The melting point of the molten salt may be adjusted by theaddition of zirconium chloride after complete removal of oxygen, andwith the amount of spent fuel added to the mix.

An additional exemplary implementation of oxidation may be furtherachieved by the addition of metal hydrides e.g., aluminum hydride orstannane (tin hydride) to enhance fast molten salt reactor nuclearproperties. Nominal density of spent fuel salt chloride is expected tobe 3.0 g/cc, depending on its (Mole %) concentration. It is anticipatedsalt fuel for the fast molten salt reactor will initially requiresignificant enrichment. This enrichment will be performed by theaddition of U235, Pu239, or MOX fuel. At an estimated beginning (30 mole%) uranium chloride and plutonium-chloride, the balance being fissionproduct chlorides and actinide chlorides (5-10) mole %, the remainingmix contains free molten salt at (60-65) mole %. Tanks 210, 220 are inexemplary implementations instrumented with dip sample points (notshown) for automatic and/or manual sampling and analysis. Thiscapability confirms independent on-line sampling that a processingtank's contents are fully mixed and chlorinated to the maximum extentpossible, substantially the entire inventory of spent nuclear fuel.

Fast reactor salt fuel requires high neutron energy for fast fission tooccur, and such energy is desired to be greater than the threshold forfast neutron energy, whereby neutrons retain enough energy after theyare produced from fission to continue the process of fast fission. Thisis achieved by conversion of spent fuel to salt fuel of heavier masselemental salt, whereby heavy mass elemental metals of potassium,zirconium, or zinc, for example, and halides of chlorine, bromine oriodine, form salt fuel effecting fast fission. Heavy mass elementalmetal hydrides of zirconium, molybdenum, or tin, for example, form saltfuel effecting fast fission by reduction of spent fuel to salt fuel andoxidation of hydride heavy mass metals to salts, whereby neutrons retainenergy well above fast neutron threshold energy after they are producedfrom fission to continue the process of fast fission.

In an exemplary implementation of the present disclosure, a method isillustrated in FIG. 8 for producing fuel for a fast molten salt reactor,the implementation of the method including:

-   -   a. providing fuel assemblies, removing fuel pellets containing        uranium and all spent fuel constituents, from the fuel        assemblies;    -   b. reducing to powder, the fuel pellets in a semi-voided        atmosphere using a ball mill, and fine mill, for process feed to        the chlorination process;    -   c. processing the granular spent salt fuel into chloride salt by        ultimate reduction and chlorination of the uranium and        associated fuel constituents chloride salt solution, by        anhydrous hydrogen chloride (AHCl); following the additions of        AHCl, salt fuel reduction may be further enhanced by addition of        metal hydrides;    -   d. enriching the spent salt fuel with U235, Pu239, or MOX;    -   e. chlorinating the enriched powdered spent salt fuel to yield        molten chloride salt fuel using AHCl halide salt reduction;    -   f. analyzing, adjusting, and certifying the molten chloride salt        fuel for end use in a molten salt reactor;    -   g. pumping the molten chloride salt fuel to stacked arrays of        cooling trays or canisters and cooling the molten chloride salt        fuel to yield solid salt fuel bars, sticks, or canister solid        form; and    -   h. milling the solidified molten chloride salt fuel to        predetermined specifications for the fast molten salt reactor.

Methods and Systems for Fluoride Fuel Salt Preparation (“Option C”)

In another exemplary implementation of the present disclosure,generally, thermal molten salt reactor fluoride salt fuel can beproduced using the same equipment apparatus (shown in FIGS. 2-6, 8, and9 ) as that used for fast molten salt reactor salt fuel in Option Bdiscussed above. As discussed above, spent fuel pellets 124 areextracted from the cladding and passed them through the ball mill 202(FIG. 9 ) and pulverized them into granular, or powder, form, wheregases are recovered from the initial disassembly, from the ball mill202, and from one or more enclosed conveyors (not shown). The powderedspent fuel is routed to the oxide reduction tanks 210 (FIG. 2 ). Thespent fuel powder is ready to be processed to either thermal or fastreactor salt fuel. Using portions of the process and an implementationof system 200 discussed above, including the removal of water from thesystem 200, recovery of gases and particulates back into the process,proceed with reduction of spent fuel constituents, uranium oxide,fission products, and actinides, into molten salt, in a generallycontinuous manner.

Thermal reactor salt fuel is prepared by addition of the powdered spentnuclear fuel to a molten salt bath of lithium and/or beryllium fluoridesalts in an oxide reduction tank 210. Quantities of fluoride molten saltcontained in the oxide reduction tanks 210, powdered spent fuel,required enrichment, and anhydrous hydrogen fluoride are determinedbefore beginning any additions. Calculations of quantities aredetermined for a specific end product Mole % of salt fuel in Mole % ofmolten salt. Partial additions of all reactants are performed withadequate time allowed for mixing and reactions, sampling andconfirmation, before further additions. Fuel salt is thoroughly mixedbefore anhydrous hydrogen fluoride (AHF) is admitted through the tanksparger arrangement 212. Salt fuel properties for a thermal salt fuelpreparation, are discussed below.

More specifically, the process begins after the spent fuel pellets 124recovered from cladding in a manner as discussed above, being fed into aball mill and fine mill 202 (FIG. 9 ), and pulverized to a powder form.Gases are recovered from the initial disassembly, from the ball mill202, and from one or more enclosed conveyors (not shown), routinggranulated spent fuel to the (FIG. 2 ) oxide reduction tanks 210. Theoxide reduction tanks which, include uranium/plutonium, are the firsttanks in line of the process to treat pulverized/powdered spent nuclearfuel. Spent fuel is reduced using a strong reducing agent, a fluoridecontaining reducing agent, such as anhydrous hydrogen fluoride (AHF)addition through a tank sparger 212 at the bottom of the tank 210.Additional reduction may be achieved by addition of metal hydrides. Asmall excess of fluoride with molten fluoride fuel salt ensures enoughfree fluoride to produce fluoride salt fuel. The reduction of uraniumoxide, plutonium and generally all spent nuclear fuel constituentsproduces hydrogen and oxygen forming water vapor and which arecontinuously removed by blower extraction and condensation.

In another exemplary implementation, numerous glow plugs (not shown)ensure hydrogen gas and oxygen are burned to water product, are placednear the top of the tank interior, have redundant power supplies andglow plug failure monitoring. All ancillary equipment for production ofwater vapor from hydrogen and oxygen, and continuous removal of waterfrom the tanks, is designed with significant margin in excess of themaximum expected process generation rate. This process completes thegoal of removing oxygen from all oxides and hydroxides in the salt fuel.Automated and dip sampling configuration, and density probes, whileprovided, are not shown. Gases are collected into a fluidized bed orsmall chemical reactor (not shown) for chlorination and recycling backinto the main process. Raw powdered spent fuel is routed from the ballmills 202 by the enclosed conveyor to parallel oxide reduction tanks 210containing molten salt. Powdered spent fuel is conveyed in a closedsystem, to the oxide reduction tank hopper 216.

A tank 220 containing molten fluoride salt maintained, in onenon-limiting example, at approximately (30-50) degrees C. (80-120degrees F.) above the melting point of the halide salt (molten alkalifluoride) melting point estimated to be 600 C (1048 F). The meltingpoint of the molten salt may be adjusted by the addition of zirconiumchloride after complete removal of oxygen, and with the amount of spentfuel added to the mix. An additional exemplary implementation ofoxidation may be further achieved by the addition of metal hydridese.g., beryllium hydride, or lithium hydride to enhance nuclearproperties for a thermal molten salt reactor. Nominal density of spentfuel salt fluoride is expected to be 3.0 g/cc, depending on its (Mole %)concentration. It is anticipated salt fuel for the thermal molten saltreactor will initially require enrichment. This enrichment will beperformed by the addition of U235, Pu239, or MOX fuel. At an estimatedbeginning (30 mole %) uranium fluoride and plutonium-fluoride, thebalance being fission product fluorides lanthanide fluorides, andactinide fluorides (5-10) mole %, the remaining mix contains free moltensalt at (60-65) mole %.

FIG. 3 shows the fuel salt mixing and adjustment tank 220, second inline of an exemplary implementation of the process, receives salt fuelin a hopper 223 from the oxide reduction tank 210. Both tanks 210, 220(FIGS. 2 and 3 ) have automated sampling, and pump recirculationdistribution headers (not shown) internal to the tanks. Tanks 210, 220(FIG. 3 ) are sized and configured to maintain subcriticality(critical-safe) in the tank as powdered spent fuel is added and enrichedwith U235, Pu239, or MOX fuel, to high-assay low enriched uranium(HALEU) at less than 20% enrichment. Both tanks 210, 220 have thecapability to receive salt, spent fuel, or enrichments; however, tank220 will normally receive only salt replenishment as needed. Theenrichment is necessitated in fueling and operation of a thermal moltensalt reactor. Tanks 210 and 220, in one non-limiting example, haveapproximate estimated dimensions of 10 feet in height by 16 feet frontto back and 10 inches wide and is capable of processing approximately600 gallons to allow the remaining free volume (head-space) forprocessing gases. Tanks 220, in one exemplary implementation, areconstructed integrally with an outside tank (not shown) having leakdetection between the inside and outside tanks. Outside tank dimensionsallow for insulation, multiple electric heater access points, recessedinstrument enclosures, and accesses to each.

The tanks 210, 220 are instrumented with dip sample points (not shown)for automatic and/or manual sampling and analysis. This capabilityconfirms independent on-line sampling that a processing tank's contentsare fully mixed and chlorinated to the extent possible, substantiallythe entire inventory of spent nuclear fuel. A density probe 221 andmanual liquid density measurement generated therefrom confirm whetherthe spent fuel salt density is at the expected density nominally(3.0-4.0) g/cm³ (kg/m³), molten alkali fluoride density, isapproximately (1.6 g/cm³). The contents of the oxide reduction tank 210(FIG. 2 ), and mixing and adjustment tank 220 will be processed furtherwhen sample analyses are confirmed. Estimated processing time is in anexemplary implementation 8 hours, including enrichment and sampleconfirmation, for one oxide reduction tank 210, and 4 hours for themixing and adjustment tank 220. The oxide reduction tanks 210 and mixingand adjustment tanks 220 are paired, may be step-wise staggered;therefore, full use would mean 4-8 hours overlap time between the firstoxide reduction tank and mixing and adjustment tank pair, and the secondoxide reduction tank and mixing and adjustment tank pair. Full rangegamma and neutron nuclear instruments, generally 224, provide continuousmonitoring, trending, and alarming (counts/second) and rate of change.In one implementation, oxide reduction tank 210 size and configurationrequire four equally spaced instruments over the height and depth ofeach tank. A blower and chiller 226 combination removes water from tank210. An anhydrous hydrogen fluoride cylinder and compressor, generally228, supply in-tank sparger arrangement 212. Salt mixers 222 are set atalternate depths, and front to back of the tank, ensure sufficientmixing of each tank. Additionally, FIG. 2 is an exemplary implementationwherein a tank transfer screw pump 218 is shown. A medium to high-volumetank screw pump 218 is connected to tank 210 via conduit 210A, and aninlet valve 218A is provided proximate the inlet of pump, and an outletvalve 218B is provided proximate the outlet of pump 218. A conduit 218Cconnects valve 218B to a discharge valve 219 connected to molten fuelsalt mixing and adjustment tank and to a tank pump recirculationisolation valve 229 for mixing oxide reduction tank contents.

In an exemplary implementation shown in FIG. 3 , a medium to high-volumetank screw pump 218 is connected to tank 220 via conduit 220A, and aninlet valve 218A is provided proximate the inlet of pump, and an outletvalve 218B is provided proximate the outlet of pump 218. A conduit 218Cconnects valve 218B to a tank header valve 217 mixing and adjustmenttank pump out connection, and to a tank pump recirculation isolationvalve 225.

In FIG. 4 , a representation of tanks 220 containing molten salt spentfuel are shown in a plan view in a side-by-side relationship, and, as inthe case of all the drawings herein, are not shown to scale. This viewrepresents the general size and configuration for both the oxidereduction tanks 210 (side by side) and mixing and adjustment tanks 220(side by side). More specifically, FIG. 4 shows an implementationwherein six tanks 220 are shown.

Accompanying tank support systems, apparatus and equipment andconfigurations used in connection with the tanks 220, are not shown.

In the basic process flow (FIG. 9 ), the first oxide reduction tank 210would pump out to the first mixing and adjustment tank, the second oxidereduction tank to the second mixing and adjustment tank, and continuethis sequence until all oxide reduction tanks have pumped out to theirrespective mixing and adjustment tanks. The tanks, FIG. 4 , also includethe tank pump discharge header and nozzles, which are only located onthe mixing and adjustment tank 220 pump out header. Spaced between tanksare encased boron slabs, or, dividers, or encasements, 230. Boronencasements 230 positioned between all oxide reduction tanks 210 andmixing and adjustment tanks 220 prevent nuclear criticality when alltanks together act as one, and maintain adequate margin ofsub-criticality communication between the array of oxide reduction tanksin close proximity, and between the array of mixing and adjustment tanksin close proximity. Each group is considered herein as one subcriticalassembly group, generally 266. Boron dividers 230 are backup defense indepth against possible criticality.

In an exemplary implementation, equipment is selected for durability andreliability. Two channels of electric “jacketed heaters” 231 (FIG. 2 )are fitted to tanks, piping, valves and pumps ensure salt fuel in pipingand equipment is of a high enough temperature to remain liquid and willflow in the event that one set of monitored heating elements fail. Theheater channels are monitored, alarmed, and component failure identifiedif such a failure occurs. If sections of piping are allowed to coolwhere molten salt is solidified, heaters can be activated to re-melt thefuel salt. Instrumentation and automated functions are fully alarmed andcontinuously monitored and displayed at the control center. Diagnosticprotocols identify and locate system failures and inform operator systemstatus interruptions or points requiring repair. All components on tanksand transfer piping must be accessible and capable of remote repairafter steps are taken to isolate failed components from the system.Multiple independent oxide reduction tanks and mixing and adjustmenttanks and transfer equipment ensures a continuous supply of fuel salt inoperation, including in the event of a system failure.

In an exemplary implementation, salt fuel preparation is begun withintroduction of fluoride salts of alkali and alkali earth metalfluorides (LiF, BeF₂), typically in crystalline form, and usually amixture of two or more salts to a tank. Heaters (electrical heatingelements) 231 are energized to melt the salt to molten state andmaintain temperature well above melting point. Pulverized powdered spentnuclear fuel is taken from the ball mill 202 and carried by enclosedconveyor to the tank hopper 216 and deposited via hopper isolation valve227 into the oxide reduction tank, and open isolation valve 217 (FIG. 2). Spent fuel addition, regulated by size and speed of the enclosedconveyor, known reaction rates derived from tests, feed limiters,ensures tank temperature is maintained within predetermined limits andsufficient mixing and reaction of tank contents occurs. All mixers andthe tank pump are turned on to initiate mixing, recirculation bypumping, and consistency of salt fuel in the tank and pump dischargelines. In an exemplary implementation, tank size and screw pump capacityare regulated to allow 4-6 hours for complete mixing and sampling,density, sample enrichment and Mole % salt fuel are recorded and asecond sample analysis completed and confirmed before a tank is ready tobe discharged.

FIGS. 5 and 6 show an exemplary implementation in plan and elevationviews of a molten salt fuel collection tray, generally 240, consistingof the tray cover 254, salt fuel collection molds 256 atop heating andcooling elements 242, in a generally checkerboard pattern of heatingelements 234A and cooling elements 234B. Insulation between elementsalong the sides and bottom of the tray prevents solidification duringpouring and minimizes cooling time after solidification. Morespecifically, FIG. 5 is a plan view of salt mold cooling tray 240 withthe tray cover and cooling molds removed.

In an exemplary implementation, salt mold cooling trays 240 (FIG. 9 )are positioned and held in a stacked array of 8-10 trays, with spacingbetween the trays being sufficient to allow for removal of the coolingmolds and cover as one assembly. Stacked arrays are tracked together bya revolving drive (not shown) which moves one stacked and cooled arraygroup, to the ball mill feed table 250. At the table 250, the one-piecemolds of each tray, in a particular group, are removed and upended oroverturned to deposit solid “bars” of salt fuel. Each tray's mold isremoved and returned to its position before the next mold is removed.After an entire array group has been emptied, it is returned in turn tobe refilled with molten salt.

An additional exemplary implementation accounts for the hygroscopicproperty of salt and salt fuel, so that each stacked array of molds isenclosed by a shroud and nitrogen inerting system (not shown) for theshort time stacked arrays are being cooled, and such stacked array andenclosed cooling system ensures cooled nitrogen is recirculated aroundthe stacked array and cooling compressor driven heat removal system.Such cooling and inerting is maintained until the stacked array solidsalt, still at high temperature, but entirely solidified, is provided tothe ball mill and fine mill, hot powdered salt fuel is put into standardcontainers or canisters, filled with argon, or cover gas and sealed.

Tray molds are a non-stick surface, with salt fuel contraction duringcooling, thereby facilitating solid salt fuel removal. The metal moldsmay be connected side-to-side and laterally supported to ensure traystrength and versatility. Solid salt “bars” are gathered to the side ofthe turning table and are generally organized lengthwise on a movingconveyor and fed into coarse ball mills 202A (FIG. 9 ). Product saltfuel from the ball mill is further conveyed to a fine mill, such as aFitz mill 252 for sizing, sampling, certification, and packaging forprotection against environmental conditions. More specifically, FIG. 6illustrates a salt mold cooling tray front view, including top cover254, cooling molds 256, heating and cooling elements 234A, 234B, which,in one implementation, could be coils.

In an alternate implementation, molten salt fuel may be stored as acontiguous solid in canisters and subcritical arrays. This processinvolves preparation of fluoride salt fuel in the aforementionedreceiving and mixing tanks 220, sampling and certification of tanks, andtransfer by screw pump to a “critical safe” standard steel canister (notshown), filled with a cover gas, and sealed, and set aside for cooling.Canisters are transported and stored, in “critical safe” arrays.Facilities using “solid salt” canisters are equipped to remotely handleand inductively heat each canister to form liquid salt fuel for additionto their molten salt reactors.

In FIG. 4 , molten salt spent fuel receiving and mixing tanks 220 areshown in a side-by-side representative top view, but not to scale. Morespecifically, FIG. 4 shows an implementation wherein sixreceiving-mixing tanks 220 are side-by-side, including a tank pumpdischarge header, generally 260, and encased boron encasements 230(sized for inner tank side dimension area). Boron encasements 230provide structural stability and protection from damage and arepositioned between salt fuel tanks 220 to maintain a sub-criticalprocess to prevent nuclear criticality communication between tanks in agiven array of tanks in close proximity, such array being consideredherein as one subcritical assembly group, generally 266. Boron dividers230 are backup defense, in depth, against possible criticality.

As shown in an exemplary implementation in FIG. 2 , the oxide reductiontank is the first tank in the process, wherein a system and process areillustrated which reduces uranium and plutonium oxides to fluorides.After this process, the contents of the reduction tank 210 are pumpedover to the mixing and adjustment tank. In one implementation, anisolation valve is provided on the loading hopper. In an exemplaryimplementation wherein only fluoride salts are desired, the salts areprepared in the oxide reduction tank, and, then pulverized, powderedspent fuel is added to the oxide reduction tank, the temperatureincrease is noted, and oxide reduction is begun by the sparging actionof anhydrous hydrogen chloride. After an allowance of time necessary formixing and water and hydrogen gas removal, toward complete reduction,the contents are pumped to the mixing and adjustment tank for finalanalysis, certification, and then pumped over to cooling trays. Theteaching of the present disclosure includes reducing all oxides andhydroxides, removing oxygen entirely and preventing production of otheroxides, ensures an authentic fluoride salt fuel, when oxygen is removedfrom the salt fuel. The result is that substantially the only byproductfrom this implementation of the present disclosure is water, which iscollected for sampling and released.

In an exemplary implementation of the present disclosure, a method isillustrated in FIG. 8 for producing fuel for a thermal molten saltreactor, the implementation of the method including:

-   -   a. providing fuel assemblies, removing fuel pellets containing        uranium and all spent fuel constituents, from the fuel        assemblies;    -   b. reducing to powder, the fuel pellets in a semi-voided        atmosphere using a ball mill, and fine mill, for process feed to        the fluorination process;    -   c. processing the powdered spent salt fuel into fluoride salt by        ultimate reduction and fluorination of the uranium and        associated fuel constituents fluoride salt solution, by        anhydrous hydrogen fluoride (AHF); following the additions of        AHF, salt fuel reduction may be further enhanced by addition of        metal hydrides;    -   d. enriching the spent salt fuel with U235, Pu239, or MOX;    -   e. fluorinating the enriched powdered spent salt fuel to yield        molten fluoride salt fuel using AHF halide salt reduction;    -   f. analyzing, adjusting, and certifying the molten fluoride salt        fuel for end use in a molten salt reactor;    -   g. pumping the molten fluoride salt fuel to stacked arrays of        cooling trays or canisters and cooling the molten fluoride salt        fuel to yield solid salt fuel bars, sticks, or canister solid        form; and    -   h. milling the solidified molten fluoride salt fuel to        predetermined specifications for the thermal molten salt        reactor.

Many modifications and other implementations of the disclosure set forthherein will come to mind to one skilled in the art to which thisdisclosure pertains having the benefit of the teachings presented in theforegoing descriptions and the associated drawings. Therefore, it is tobe understood that the disclosure is not to be limited to the specificimplementations disclosed and that modifications and otherimplementations are intended to be included within the scope of theappended claims.

Moreover, although the foregoing descriptions and the associateddrawings describe example implementations in the context of certainexample combinations of elements and/or functions, it should beappreciated that different combinations of elements and/or functions maybe provided by alternative implementations without departing from thescope of the appended claims. In this regard, for example, differentcombinations of elements and/or functions than those explicitlydescribed above are also contemplated as may be set forth in some of theappended claims. Although specific terms are employed herein, they areused in a generic and descriptive sense only and not for purposes oflimitation.

1. A method for use in processing spent nuclear fuel having uranium intomolten salt reactor fuel, the method comprising: feeding the spentnuclear fuel to: a halide forming process, wherein the halide includesat least one of chloride, bromide, and iodide, and processing the spentnuclear fuel into halide salt by ultimate reduction by reacting thehalide salt with at least one of anhydrous hydrogen halide and metalhydride in an oxide reduction tank; and halide forming of the uraniumand associated fuel constituents in a halide salt solution comprised ofa bath of selected metal hydride salts; or a fluoride forming process,and processing the spent fuel into fluoride salt by ultimate reduction;and fluoride forming of the uranium and associated fuel constituents ina fluoride salt solution comprised of a bath of selected metal hydridesalts.
 2. A method for use in processing spent nuclear fuel havinguranium into molten salt reactor fuel, the method comprising: feedingthe spent nuclear fuel to a halide forming process, wherein the halideincludes at least one of chloride, bromide, and iodide, and processingthe spent nuclear fuel into halide salt by ultimate reduction byreacting the halide salt with at least one of anhydrous hydrogen halideand metal hydride in an oxide reduction tank; and halide forming of theuranium and associated fuel constituents in a halide salt solutioncomprised of a bath of selected metal hydride salts.
 3. The method ofclaim 2, further comprising enriching the halide salt in the oxidereduction tank and halogenating the enriched halide salt to yield halidesalt fuel.
 4. The method of claim 2, wherein the step of processing thespent nuclear fuel into halide salt occurs by reacting the halide saltwith at least one of anhydrous hydrogen halide and metal hydride.
 5. Themethod of claim 2, wherein the step of processing the spent nuclear fuelinto halide salt occurs by reacting the halide salt with at least one ofanhydrous hydrogen halide and metal hydride.
 6. A method for use inprocessing spent nuclear fuel having uranium into molten salt reactorfuel, the method comprising: feeding the spent nuclear fuel to afluoride forming process; processing the spent nuclear fuel intofluoride salt by ultimate reduction; and fluoride forming of the uraniumand associated spent nuclear fuel constituents in a fluoride saltsolution comprised of a bath of selected metal hydride salts.
 7. Themethod of claim 6, further comprising enriching the fluoride salt in anoxide reduction tank.
 8. The method of claim 7, further comprisingfluorinating the enriched fluoride salt to yield molten fluoride saltfuel.
 9. The method of claim 6, wherein the step of processing the spentnuclear fuel into fluoride salt occurs by reacting the fluoride saltwith anhydrous hydrogen fluoride.
 10. The method of claim 6, wherein thestep of processing the spent nuclear fuel into fluoride salt occurs byreacting the fluoride salt with anhydrous hydrogen fluoride in an oxidereduction tank.
 11. The method of claim 6, wherein the step ofprocessing the spent nuclear fuel into fluoride step occurs by reactingthe fluoride salt with anhydrous hydrogen fluoride via a sparger in anoxide reduction tank.
 12. A method for use in processing uranium intomolten salt reactor fuel, the method comprising: feeding the uranium toa halide forming process, wherein the halide includes at least one ofchloride, bromide, and iodide, and processing the uranium into halidesalt by ultimate reduction by reacting the halide salt with at least oneof anhydrous hydrogen halide and metal hydride in an oxide reductiontank; and halide forming of the uranium in a halide salt solutioncomprised of a bath of selected metal hydride salts.
 13. A method foruse in processing uranium into molten salt reactor fuel, the methodcomprising: processing the uranium into fluoride salt by ultimatereduction; and fluoride forming of the uranium in a fluoride saltsolution comprised of a bath of selected metal hydride salts.
 14. Asystem for use in processing uranium into molten salt reactor fuel, thesystem comprising: means for feeding uranium to a halide formingprocess, wherein the halide includes at least one of chloride, bromide,and iodide, and processing the uranium into halide salt by ultimatereduction by reacting the halide salt with at least one of anhydroushydrogen halide and metal hydride in an oxide reduction tank; and meansfor halide forming of the uranium in a halide salt solution comprised ofa bath of selected metal hydride salts.
 15. A system for use inprocessing uranium into molten salt reactor fuel, the system comprising:means for processing the uranium into fluoride salt by ultimatereduction; and means for fluoride forming of the uranium in a fluoridesalt solution comprised of a bath of selected metal hydride salts.
 16. Amethod for use in processing spent nuclear fuel having uranium, fissionproduct waste, and actinide constituents into molten salt reactor fuel,the method comprising: immersing the spent nuclear fuel: in a moltenfluoride salt bath and converting the spent nuclear fuel into a fluoridesalt fuel for a molten salt reactor, without chemically separating thefission product waste and actinide constituents from the spent nuclearfuel; or in a molten chloride salt bath and converting the spent nuclearfuel into a chloride salt fuel for a molten salt reactor, withoutchemically separating the fission product waste and actinideconstituents from the spent nuclear fuel.
 17. The method claim 16,wherein the converting of the spent nuclear fuel into a fluoride saltfuel does not include aqueous wet chemical separation.
 18. The methodclaim 16, wherein the converting of the spent nuclear fuel into achloride salt fuel does not include aqueous wet chemical separation. 19.A method for use in processing spent nuclear fuel having uranium,fission product waste, and actinide constituents into molten saltreactor fuel, the method comprising: immersing the spent nuclear fuel ina molten fluoride salt bath; and converting the spent nuclear fuelimmersed in the molten chloride salt bath into a fluoride salt fuel fora molten salt reactor, without chemically separating the fission productwaste and actinide constituents from the spent nuclear fuel.
 20. Amethod for use in processing spent nuclear fuel having uranium, fissionproduct waste, and actinide constituents into molten salt reactor fuel,the method comprising: immersing the spent nuclear fuel in a moltenchloride salt bath; and converting the spent nuclear fuel immersed inthe molten chloride salt bath into a chloride salt fuel for a moltensalt reactor, without chemically separating the fission product wasteand actinide constituents from the spent nuclear fuel.
 21. A system foruse in processing spent nuclear fuel having uranium, fission productwaste, and actinide constituents into molten salt reactor fuel, thesystem comprising: a molten fluoride salt bath; and means for immersingthe spent nuclear fuel in the molten fluoride salt bath and forconverting the spent nuclear fuel into a fluoride salt fuel for a moltensalt reactor, without chemically separating the fission product wasteand actinide constituents from the spent nuclear fuel.
 22. A system foruse in processing spent nuclear fuel having uranium, fission productwaste, and actinide constituents into molten salt reactor fuel, thesystem comprising: a molten chloride salt bath; and means for immersingthe spent nuclear fuel in the molten chloride salt bath and forconverting the spent nuclear fuel into a chloride salt fuel for a moltensalt reactor, without chemically separating the fission product wasteand actinide constituents from the spent nuclear fuel.